Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 14303-14312 [E7-5342]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 58 (Tuesday, March 27, 2007)]
[Notices]
[Pages 14303-14312]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-5342]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 2, 2007 to March 15, 2007. The last 
biweekly notice was published on March 13, 2007 (72 FR 11383).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, Division of Administrative Services, 
Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area O1F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should

[[Page 14304]]

consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc.

    Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi.
    Date of amendment request: February 8, 2007.
    Description of amendment request: The proposed amendment would 
modify Grand Gulf Nuclear Station, Unit 1 (GGNS) technical 
specification (TS) requirements for MODE change limitations in limiting 
condition for operation (LCO) 3.0.4 and surveillance requirement (SR) 
3.0.4. The proposed TS changes are consistent with Revision 9 of 
Nuclear Regulatory Commission (NRC) approved Industry TS Task Force 
(TSTF) Standard TS Change Traveler, TSTF-359, ``Increase Flexibility in 
MODE Restraints.'' In addition, the proposed amendment would also 
change TS Section 1.4, Frequency, Example 1.4-1, ``Surveillance 
Requirements,'' to accurately reflect the changes made by TSTF-359, 
which is consistent with NRC-approved TSTF-485, Revision 0, ``Correct 
Example 1.4-1.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 2, 2002 (67 FR 50475), as part of the 
Consolidated Line Item Improvement Process (CLIIP), on possible 
amendments to revise the

[[Page 14305]]

plant-specific TS to modify requirements for MODE change limitations in 
LCO 3.0.4 and SR 3.0.4.
    The NRC staff subsequently issued a notice of availability of the 
models for Safety Evaluation and No Significant Hazards Consideration 
Determination for referencing in license amendment applications in the 
Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed 
the applicability of the CLIIP, including the model No Significant 
Hazards Consideration Determination, in its application dated February 
8, 2007.
    The proposed TS changes are consistent with NRC-approved Industry 
TSTF Standard TS change, TSTF-359, Revision 8, as modified by 68 FR 
16579. TSTF-359, Revision 8, was subsequently revised to incorporate 
the modifications discussed in the April 4, 2003, Federal Register 
notice and other minor changes. TSTF-359, Revision 9, was subsequently 
submitted to the NRC on April 28, 2003, and was approved by the NRC on 
May 9, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the NRC staff's analysis 
of the issue of no significant hazards consideration is presented 
below:
Criterion 1--The Proposed Changes Do Not Involve a Significant Increase 
in the Probability or Consequences of an Accident Previously Evaluated
    The proposed changes in TS Section 1.4, Frequency, Example 1.4-1, 
would accurately reflect the changes made by TSTF-359 in LCO 3.0.4 and 
SR 3.0.4, which are consistent with NRC-approved TSTF-485, Revision 0. 
These changes are considered administrative in that they modify the 
example to demonstrate the proper application of LCO 3.0.4 and SR 
3.0.4. The requirements of LCO 3.0.4 and SR 3.0.4 are clear and are 
clearly explained in the associated Bases. As a result, modifying the 
example will not result in a change in usage of the TS.
    The proposed changes in LCO 3.0.4 and SR 3.0.4 allow entry into a 
mode or other specified condition in the applicability of a TS, while 
in a TS condition statement and the associated required actions of the 
TS. The proposed changes do not adversely affect accident initiators or 
precursors, the ability of structures, systems, and components to 
perform their intended function to mitigate the consequences of an 
initiating event within the assumed acceptance limits, or radiological 
release assumptions used in evaluating the radiological consequences of 
an accident previously evaluated. Being in a TS condition and the 
associated required actions are not an initiator of any accident 
previously evaluated. Therefore, the probability of an accident 
previously evaluated is not significantly increased. The consequences 
of an accident while relying on required actions as allowed by proposed 
LCO 3.0.4, are no different than the consequences of an accident while 
entering and relying on the required actions while starting in a 
condition of applicability of the TS. Therefore, the consequences of an 
accident previously evaluated are not significantly affected by these 
changes. The addition of a requirement to assess and manage the risk 
introduced by these changes will further minimize possible concerns. 
Therefore, these changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Changes Do Not Create the Possibility of a 
New or Different Kind of Accident from any Previously Evaluated
    No new or different accidents result from utilizing the proposed 
changes. The proposed changes do not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) or 
a change in the methods governing normal plant operation. In addition, 
the changes do not impose any new or different requirements or 
eliminate any existing requirements. The proposed changes do not alter 
assumptions made in the safety analysis and are consistent with the 
safety analysis assumptions and current plant operating practice. 
Entering into a mode or other specified condition in the applicability 
of a TS, while in a TS condition statement and the associated required 
actions of the TS, will not introduce new failure modes or effects and 
will not, in the absence of other unrelated failures, lead to an 
accident whose consequences exceed the consequences of accidents 
previously evaluated. The addition of a requirement to assess and 
manage the risk introduced by these changes will further minimize 
possible concerns. Thus, these changes do not create the possibility of 
a new or different kind of accident from an accident previously 
evaluated.
Criterion 3--The Proposed Changes Do Not Involve a Significant 
Reduction in the Margin of Safety
    The proposed changes in TS Section 1.4, Example 1.4-1, are 
considered administrative and will have no effect on the application of 
the TS requirements. Therefore, the margin of safety provided by the TS 
requirements is unchanged. The proposed changes in TS LCO 3.0.4 and SR 
3.0.4 allow entry into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS. The GGNS TS allows operation of 
the plant without the full complement of equipment through the TS 
conditions for not meeting the TS LCO. The risk associated with this 
allowance is managed by the imposition of required actions that must be 
performed within the prescribed completion times. The net effect of 
being in a TS LCO condition on the margin of safety is not considered 
significant. The proposed changes do not alter the required actions or 
completion times of the TS. The proposed changes allow TS conditions to 
be entered, and the associated required actions and completion times to 
be used in new circumstances. This use is predicated upon the 
licensee's performance of a risk assessment and the management of plant 
risk. The changes also eliminate current allowances for utilizing 
required actions and completion times in similar circumstances, without 
assessing and managing risk. The net change to the margin of safety is 
insignificant. Therefore, these changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: David Terao.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 
and 2, Will County, Illinois.
    Date of amendment request: January 8, 2007.
    Description of amendment request: The proposed amendment would 
revise the technical specification (TS) requirements for selected 
reactor trip system (RTS) instrumentation, engineered safety feature 
actuation system (ESFAS) instrumentation, and containment ventilation 
isolation instrumentation to adopt completion times, test bypass time, 
and surveillance test interval changes. The changes are based on 
Westinghouse Electric Company, LLC, topical reports WCAP-14333-P-A, 
Revision 1, ``Probabilistic

[[Page 14306]]

Risk Analysis of the [Reactor Protection System] RPS and ESFAS Test 
Times and Completion Times,'' and WCAP-15376-P-A, Revision 1, ``Risk-
Informed Assessment of the RTS and ESFAS Surveillance Test Intervals 
and Reactor Trip Breaker Test and Completion Times.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since no 
hardware changes are proposed. The same RTS and ESFAS 
instrumentation will continue to be used. The protection systems 
will continue to function in a manner consistent with the plant 
design basis. These changes to the TS do not result in a condition 
where the design, material, and construction standards that were 
applicable prior to the change are altered.
    The proposed changes will not modify any system interface. The 
proposed changes will not affect the probability of any event 
initiators. There will be no degradation in the performance of or an 
increase in the number of challenges imposed on safety-related 
equipment assumed to function during an accident situation. There 
will be no change to normal plant operating parameters or accident 
mitigation performance. The proposed changes will not alter any 
assumptions or change any mitigation actions in the radiological 
consequence evaluations in the Updated Final Safety Analysis Report.
    The determination that the results of the proposed changes are 
acceptable was established in the NRC Safety Evaluations prepared 
for WCAP-14333-P-A, (issued by letter dated July 15, 1998) and for 
WCAP-15376-P-A, (issued by letter dated December 20, 2002). 
Implementation of the proposed changes will result in an 
insignificant risk impact.
    Applicability of these conclusions has been verified through 
plant-specific reviews and implementation of the generic analysis 
results in accordance with the respective NRC Safety Evaluation 
conditions.
    The proposed changes to the CTs [completion times], test bypass 
times, and Surveillance Frequencies reduce the potential for 
inadvertent reactor trips and spurious engineered safeguard features 
actuations, and therefore do not increase the probability of any 
accident previously evaluated. The proposed changes do not change 
the response of the plant to any accidents and have an insignificant 
impact on the reliability of the RTS and ESFAS signals. The RTS and 
ESFAS will remain highly reliable and the proposed changes will not 
result in a significant increase in the risk of plant operation. 
This is demonstrated by showing that the impact on plant safety, as 
measured by the increase in core damage frequency (CDF) is less than 
1.0E-06 per year and the increase in large early release frequency 
(LERF) is less than 1.0E-07 per year. In addition, for the CT 
changes, the incremental conditional core damage probabilities 
(ICCDP) and incremental conditional large early release 
probabilities (ICLERP) are less than 5.0E-07 and 5.0E-08, 
respectively. These changes meet the acceptance criteria in 
Regulatory Guides (RGs) 1.174 and 1.177. Therefore, since the RTS 
and ESFAS will continue to perform their functions with high 
reliability, as originally assumed, and the increase in risk, as 
measured by [Delta]CDF, [Delta]LERF, ICCDP, ICLERP risk metrics, is 
within the acceptance criteria of existing regulatory guidance, 
there will not be a significant increase in the consequences of any 
accidents.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components from 
performing their intended function to mitigate the consequences of 
an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of any accident previously evaluated. 
The proposed changes are consistent with safety analysis assumptions 
and resultant consequences.
    Therefore, this change does not increase the probability or 
consequences of any accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. There are no hardware changes nor are there any changes 
in the method by which any safety-related plant system performs its 
safety function. The proposed changes will not affect the normal 
method of plant operation. No performance requirements will be 
affected or eliminated. The proposed changes will not result in 
physical alteration to any plant system nor will there be any change 
in the method by which any safety-related plant system performs its 
safety function. There will be no setpoint changes or changes to 
accident analysis assumptions.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of these changes. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of these changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety?
    The proposed changes do not affect the acceptance criteria for 
any analyzed event nor is there a change to any Safety Analysis 
Limit. There will be no effect on the manner in which safety limits, 
limiting safety system settings, or limiting conditions for 
operation are determined nor will there be any effect on those plant 
systems necessary to assure the accomplishment of protection 
functions. There will be no impact on the departure from nucleate 
boiling limits, fuel centerline temperature, or any other margin of 
safety. The radiological dose consequence acceptance criteria listed 
in the NUREG-0800, ``Standard Review Plan for the Review of Safety 
Analysis Reports for Nuclear Power Plants,'' will continue to be 
met.
    Redundant RTS and ESFAS trains are maintained, and diversity 
with regard of the signals that provide reactor trip and engineered 
safety features actuation is also maintained. All signals credited 
as primary or secondary, and all operator actions credited in the 
accident analyses will remain the same. The proposed changes will 
not result in plant operation in a configuration outside the design 
basis. The calculated impact on risk is insignificant and meets the 
acceptance criteria contained in RGs 1.174 and 1.177. Although there 
was no attempt to quantify any positive human factors benefit due to 
increased CTs and bypass test times, it is expected that there would 
be a net benefit due to a reduced potential for spurious reactor 
trips and actuations associated with testing.
    Implementation of the proposed changes is expected to result in 
an overall improvement in safety, as follows:
     Reduced testing will result in fewer inadvertent 
reactor trips, less frequent actuation of ESFAS components, less 
frequent distraction of operations personnel without significantly 
affecting RTS and ESFAS reliability.
     Improvements in the effectiveness of the operating 
staff in monitoring and controlling plant operation will be 
realized. This is due to less frequent distraction of the operators 
and shift supervisor to attend to instrumentation Required Actions 
with short CTs.
     Longer repair times associated with increased CTs will 
lead to higher quality repairs and improved reliability.
     The CT extensions for the reactor trip breakers will 
provide additional time to complete test and maintenance activities 
while at power, potentially reducing the number of forced outages 
related to compliance with reactor trip breaker CT, and provide 
consistency with the CT for the logic trains.

    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell A. Gibbs.

[[Page 14307]]

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: January 30, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) Surveillance Requirement (SR) 
3.5.1.3.b to correctly state that the required pressure at which the 
Alternate Nitrogen System is determined to be operable should be 
greater than or equal to 410 psig, not the currently stated pressure of 
greater than or equal to 220 psig. The safety-related Alternate 
Nitrogen System provides an alternate pressure source to equipment 
required during or following an accident. The licensee has determined 
that the current acceptance value specified by SR 3.5.1.3.b is non-
conservative and needs to be corrected to the higher value.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Part 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration (NSHC). 
The NRC staff reviewed the licensee's analysis, and has performed its 
own analysis as follows:

    (1) Does the proposed amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    No. The proposed amendment would only correct the acceptance 
value specified by SR 3.5.1.3.b. The acceptance value of the 
nitrogen supply was not considered to be a precursor to, and does 
not affect the probability of, an accident. In addition, there is no 
design or operation change associated with the proposed amendment.
    Therefore, the proposed amendment does not increase the 
probability of an accident previously evaluated.
    The corrected, higher pressure of the Alternate Nitrogen System 
will ensure that nitrogen is available to operate equipment after an 
accident, as designed. The increased acceptance value will not 
decrease the functionality of the Alternate Nitrogen System, or the 
functionality of the plant equipment it supports. Therefore, the 
plant systems required to mitigate accidents will remain capable of 
performing their design functions. As a result, the proposed 
amendment will not lead to a significant change in the consequences 
of any accident.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed amendment does not involve a physical 
alteration of any system, structure, or component (SSC) or a change 
in the way any SSC is operated. The proposed amendment does not 
involve operation of any SSCs in a manner or configuration different 
from those previously recognized or evaluated. No new failure 
mechanisms will be introduced by the revised acceptance value.
    Thus, the proposed amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed amendment only changes the acceptance value of 
the Alternate Nitrogen System. There will be no modification of any 
TSs limiting condition for operation, no change to any limit on 
previously analyzed accidents, no change to how previously analyzed 
accidents or transients would be mitigated, no change in any 
methodology used to evaluate consequences of accidents, and no 
change in any operating procedure or process. Therefore, the 
proposed amendment does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
the NRC staff's own analysis above, it appears that the three standards 
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the proposed amendment involves no significant hazards 
consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Branch Chief: L. Raghavan.

Southern California Edison Company, et al. Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of amendment requests: February 8, 2007.
    Description of amendment requests: This license amendment request 
will (1) revise Technical Specification (TS) Surveillance Requirement 
(SR) 3.3.7.3.a to lower the allowable value for dropout and raise the 
allowable value for pickup of the degraded voltage function, and (2) 
revise TS SR 3.8.1 to lower the diesel generator minimum output voltage 
due to lower settings for the degraded voltage function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This proposed change revises the Technical Specification (TS) 
Surveillance Requirement (SR) 3.3.7.3.a allowable values of the 
Degraded Voltage Function and SRs 3.8.1.2, .7, .9, .11, .12, .15, 
.16, .17, .19, and .20 for Diesel Generator (DG) minimum operable 
voltage. This proposed change will allow Southern California Edison 
(SCE) to widen the operating band while maintaining adequate 
conservatism for the degraded relay settings and overall loop 
uncertainties while keeping 218 kV as the minimum voltage on the 
offsite transmission grid necessary to support operability of the 
immediate access offsite power source (also referred to as the 
normal preferred power source). This will be accomplished by 
lowering the dropout and increasing the pickup settings of the 
degraded voltage protection relays. Following approval of this 
proposed change, the 4.16 kV Class 1E buses would remain on the 
normal preferred power source at or above a grid voltage of 218 kV 
while protecting all Class 1E equipment from degraded grid 
conditions.
    The degraded voltage protection circuits are designed to protect 
electrical equipment against the effects of degraded voltage on the 
offsite transmission networks. Therefore, these circuits are 
generally not considered to be accident initiators. However, 
spurious actuation of the degraded voltage protection relays could 
result in the loss of the preferred power source (offsite source of 
alternating current (AC) power). The proposed change lowers the 
allowable value for dropout and raises the allowable value for 
pickup for the degraded voltage protection relays. This results in 
an increase in operating band and a lower probability of spurious 
actuation of these degraded voltage signals. Therefore, there is no 
increase in the probability of a Loss of Offsite Power (preferred 
power source) as a result of this proposed change.
    The safety function of the degraded voltage protection circuits 
is to ensure the operability of Class 1E equipment. SCE has 
performed calculations that demonstrate that operation in accordance 
with this proposed change will not result in operation of plant 
equipment at degraded voltages. Therefore, there is no increase in 
the consequences of any accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed allowable values of the degraded voltage relays and 
the DG minimum operating voltage will provide an acceptable level of 
protection for plant equipment.
    This proposed change affects only the voltage settings of the 
degraded voltage protection relays and voltage regulator setting of 
the DG for lowering the required bus voltage. There is no other 
change to the degraded voltage function. There are no physical 
modifications necessary to the degraded voltage protection relays or 
the DG. There are no changes to the actions performed by the relays 
or the DG following actuation. Therefore, there are no new failure

[[Page 14308]]

modes or effects introduced by this proposed change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed degraded voltage protection schemes are designed to 
ensure that plant equipment will not operate at a degraded voltage 
and the DG Automatic Voltage Regulator (AVR) is set to provide 
adequate voltage for resetting of the relays and satisfactory 
operation of the Safety Related equipment. The proposed degraded 
voltage allowable values will not affect the existing protection 
criterion for plant equipment. This maintains the existing margin of 
safety for plant equipment.
    Therefore, there is no significant reduction in a margin of 
safety as a result of the proposed amendment.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: David Terao.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia

    Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 
1 and 2, Appling County, Georgia.
    Date of amendment request: February 13, 2007.
    Description of amendment request: The proposed amendment would 
modify the licensee's Technical Specification (TS) Section 3.9.1, 
``Refueling Equipment Interlocks,'' to add required actions to allow 
insertion of a control rod withdrawal block and verification that all 
control rods are fully inserted as alternate actions to suspending in-
vessel fuel movement in the event that one or more required refueling 
equipment interlocks are inoperable. These changes are based on 
Technical Specification Task Force (TSTF) change TSTF-225, Revision 2, 
``Fuel movement with inoperable refueling equipment interlocks'' and 
are consistent with the current Boiling Water Reactor (BWR)/4 Standard 
Technical Specifications (STS), NUREG-1433, Volume 1, Revision 3.0.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change provides additional actions for an 
inoperable refueling equipment interlock. The proposed actions will 
allow fuel movement with inoperable refueling interlocks, however, 
those actions will require the insertion of a continuous control rod 
withdrawal block, as well as verification that all control rods are 
fully inserted, before the commencement of fuel movement. Since fuel 
movement with the refueling interlocks operable allows control rod 
withdrawal under some circumstances, complete prevention of control 
rod withdrawal with the refueling interlocks inoperable does not 
increase the likelihood of a reactivity event, and may in fact 
decrease its probability of occurrence.
    The refueling interlocks are not designed or otherwise intended 
to prevent or mitigate the consequences of the fuel handling 
accident. This proposed change does not involve those structures 
that could have an effect on the fuel handling accident and its 
consequences, such as the fuel design, the integrity of the 
refueling platform, and the integrity of the refueling mast and 
grapple. Furthermore, the consequences of the refueling accident are 
not increased since, should that accident occur while operating 
under the provisions of the alternate actions, all control rods will 
be fully inserted. The consequences of the fuel assembly insertion 
error event during refueling are not increased since this proposed 
change preserves the initial conditions of that transient event, 
i.e., all control rods inserted.
    Implementing these changes will not increase the likelihood of 
an equipment failure resulting from the use of the refueling cranes 
and hoists. Such protection is afforded by other plant (owner 
controlled) specifications and procedures. These documents require 
testing and maintenance of these components separate from the 
requirements of [Limiting Condition for Operation] LCO 3.9.1.
    This submittal does not affect any other system, structure or 
component that is important with respect to the prevention and 
mitigation of other accidents or transients.
    For the above reasons, this proposed Technical Specifications 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change provides additional actions (the insertion 
of a control rod block and verification that all control rods are 
fully inserted) for inoperable refueling interlocks. This change 
does not involve any permanent alterations to plant systems or 
components. Nor does it involve changes to operational 
configurations or to the maintenance and testing of systems or 
components. Consequently, no new modes of operation are being 
introduced. Therefore, the change does not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    The proposed change provides additional actions for an 
inoperable required refueling equipment interlock. The new actions 
will require that all control rods be fully inserted and that a 
control rod block be in effect. Under the current specifications, 
control rod withdrawal is allowed during fuel movement under certain 
conditions.
    The alternate actions of the proposed specifications will not 
allow rod withdrawal under any circumstances during fuel movement 
operations, therefore, this proposed change provides a level of 
safety at least equivalent to the existing actions.
    Consequently, the change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Branch Chief: Evangelos C. Marinos.

Virginia Electric and Power Company

    Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 
2, Surry County, Virginia.
    Date of amendment request: February 26, 2007.
    Description of amendment request: The proposed change adds an 
operating license condition and revises the Technical Specifications to 
permit the replacement of main control room (MCR) and emergency 
switchgear room (ESGR) air-conditioning system (ACS) chilled water 
piping by using temporary 45-day and 14-day allowed outage times (AOTs) 
four times in a 24-month time span.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed change has been evaluated using the risk-informed 
processes described

[[Page 14309]]

in Regulatory Guide (RG) 1.174, ``An Approach for Using 
Probabilistic Risk assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' and RG 1.177, ``An 
Approach for Plant-Specific, Risk-Informed Decision Making: 
Technical Specifications.''
    The risk associated with the proposed change was found to be 
acceptably ``small'' and therefore not a significant increase in the 
probability or consequences of an accident previously evaluated.
    In addition, the proposed change does not affect the initiators 
of analyzed events or the assumed mitigation of accident or 
transient events. During the temporary 45-day and 14-day AOT 
entries, equipment availability restrictions will restrict or limit 
the out-of-service time of risk significant plant equipment due to 
surveillance testing, preventive maintenance, and elective 
maintenance. In addition, during the replacement activities, 
compensatory actions will be in place to ensure the availability of 
chilled water or to provide backup cooling. Therefore, the ACS will 
continue to perform its required function. As a result, the proposed 
change to the Surry TS does not involve any significant increase in 
the probability or the consequences of any accident or malfunction 
of equipment important to safety previously evaluated since neither 
accident probabilities nor consequences are being affected by this 
proposed change.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    The proposed change does not involve a change in the methods 
used to respond to plant transients. There is no alteration to the 
parameters within which the plant is normally operated or in the 
setpoints, which initiate protective or mitigative actions. The MCR 
and ESGR ACS will continue to perform its required function. This is 
assured by the planned implementation of compensatory actions, 
including provisions for backup cooling. Consequently, no new 
failure modes are introduced by the proposed change. Therefore, the 
proposed Surry TS change does not create the possibility of a new or 
different kind of accident or malfunction of equipment important to 
safety from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Margin of safety is established through the design of the plant 
structures, systems, and components, the parameters within which the 
plant is operated, and the establishment of the setpoints for the 
actuation of equipment relied upon to respond to an accident or 
transient event. The proposed change does not affect the ability of 
the MCR and ESGR ACS to perform its required function. This is 
assured by the planned implementation of compensatory actions, 
including provisions for backup cooling. Furthermore, the proposed 
change has been evaluated using the risk-informed processes 
described in Regulatory Guide (RG) 1.174, ``An approach for Using 
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' and RG 1.177, ``An 
Approach for Plant-Specific, Risk-Informed Decision Making: 
Technical Specifications.''
    The risk associated with the proposed change was found to be 
acceptably small. Therefore, the proposed change to the Surry TS 
does not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Branch Chief: Evangelos C. Marinos.

Virginia Electric and Power Company

    Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 
2, Surry County, Virginia.
    Date of amendment request: March 6, 2007.
    Description of amendment request: The proposed amendments would 
revise the licensing basis (Updated Final Safety Analysis Report 
(UFSAR)) to permit irradiation of the fuel assemblies beginning with 
Surry Power Station, Unit Nos. 1 and 2, improved fuel assemblies with 
ZIRLO (Westinghouse trademark) cladding to a lead rod average burnup of 
62,000 MWD/MTU.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The probability of occurrence or the consequence of an 
accident previously evaluated is not significantly increased.
    The activity being evaluated is a slight increase in the lead 
rod average burnup limit for the fuel assemblies. No change in fuel 
design or fuel enrichment will be required to increase the lead rod 
average burnup. The fuel rods at the extended lead rod average 
burnup will continue to meet the design limits with respect to fuel 
rod growth, clad fatigue, rod internal pressure and corrosion. There 
will be no impact on the capability to engage the fuel assemblies 
with the handling tools. Therefore, it is concluded that the change 
will not result in an increase in the probability of occurrence of 
any accident previously evaluated in the UFSAR. The impact of 
extending the lead rod average burnup to 62,000 MWD/MTU from 60,000 
MWD/MTU on the core kinetics parameter, core thermal-hydraulics/
[departure from nucleate boiling ratio]DNBR, specific accident 
considerations, and radiological consequences was considered. Based 
on the evaluation of these considerations, it is concluded that 
increasing the lead rod average burnup limit to 62,000 MWD/MTU will 
not result in a significant increase in the consequences of the 
accidents previously evaluated in the Surry UFSAR.
    2. The possibility for a new or different type of accident from 
any accident previously evaluated is not created.
    The fuel is the only component affected by the change in the 
burnup limit. The change does not affect the thermal hydraulic 
response to any transient or accident. The existing fuel rod design 
criteria continue to be met at the higher burnup limit. Thus, the 
change does not create the possibility of an accident of a different 
type.
    3. The margin of safety as defined in the Bases to the Surry 
Technical Specifications is not significantly reduced.
    The operation of the Surry cores with a limited number of fuel 
assemblies with some fuel rods irradiated to a lead rod average 
burnup of 62,000 MWD/MTU will not change the performance 
requirements of any system or component such that any design 
criteria will be exceeded. The normal limits on core operation 
defined in the Surry Technical Specifications will remain applicable 
for the irradiation of the fuel to a lead rod average burnup of 
62,000 MWD/MTU. Therefore, the margin of safety as defined in the 
Bases to the Surry Technical Specifications is not significantly 
reduced.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Branch Chief: Evangelos C. Marinos.

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.

[[Page 14310]]

    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Florida Power Corporation, et al.

    Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant, 
Citrus County, Florida.
    Date of amendment request: February 8, 2007.
    Description of amendment request: To change the basis for 
protection of spent fuel stored in the spent fuel pool (SFP) in order 
to eliminate the Final Safety Analysis Report commitment for 
maintaining the SFP missile shields.
    Date of publication of individual notice in the Federal Register: 
March 13, 2007. (72 FR 11381).
    Expiration date of individual notice: May 14, 2007.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments.
    If the Commission has prepared an environmental assessment under 
the special circumstances provision in 10 CFR 51.22(b) and has made a 
determination based on that assessment, it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

Carolina Power & Light Company

    Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2, 
Darlington County, South Carolina.
    Date of application for amendment: May 30, 2006, as supplemented by 
letter dated November 20, 2006.
    Brief description of amendment: The amendment revises the existing 
steam generator tube surveillance program at H. B. Robinson Steam 
Electric Plant, Unit No. 2.
    Date of issuance: March 12, 2007.
    Effective date: This license amendment is effective as of the date 
of issuance and shall be implemented within 60 days.
    Amendment No. 212.
    Renewed Facility Operating License No. DPR-23. Amendment revises 
the Technical Specifications.
    Date of initial notice in Federal Register: December 19, 2007 (71 
FR 75990). The November 20, 2006, supplemental letter provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated: March 12, 2007.
    No significant hazards consideration comments received: No.

Dominion Energy Kewaunee, Inc.

    Docket No. 50-305, Kewaunee Power Station, Kewaunee County, 
Wisconsin.
    Date of application for amendment: January 30, 2006, as 
supplemented by letter dated January 23, 2007.
    Brief description of amendment: The amendment modifies the 
radiological accident analyses and associated technical specifications.
    Date of issuance: March 8, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 190.
    Facility Operating License No. DPR-43: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 14, 2006 (71 FR 
13172).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination, and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 8, 2007.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc.

    Docket No. 50-423, Millstone Power Station, Unit No 3, New London 
County, Connecticut.
    Date of application for amendment: February 7, 2006, as 
supplemented by letters dated August 14, 2006, and January 2, 2007.
    Brief description of amendments: The amendment revised the 
Millstone Power Station, Unit No. 3 Technical Specifications to permit 
an increase in the allowed outage time from 72 hours to 7 days for the 
inoperablity of the steam supply to the turbine-driven auxiliary 
feedwater pump (AFW) or the inoperability of the turbine-driven AFW 
pump under certain operating mode restrictions.
    Date of issuance: February 28, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 235.
    Facility Operating License No NPF-49: Amendment revised the License 
and Technical Specifications.
    Date of initial notice in Federal Register: April 11, 2006 (70 FR 
18372).
    The supplements dated August 14, 2006, and January 2, 2007, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 28, 2007.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc.

    Docket Nos. 50-313 and 50-368, Arkansas Nuclear One, Units 1 and 2, 
Pope County, Arkansas.
    Date of amendment request: October 25, 2005, as supplemented by 
letter dated March 20, 2006.

[[Page 14311]]

    Brief description of amendments: The changes addressed inventory 
and inspection requirements associated with the emergency cooling pond, 
which is a common cooling water source for both units during conditions 
that may render the normal cooling water source (Dardanelle Reservoir) 
unavailable.
    Date of issuance: March 9, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: Unit 1-229, Unit 2-271.
    Renewed Facility Operating License Nos. DPR-51 and NPF-6: 
Amendments revised the Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: October 24, 2006 (71 FR 
62309). The supplemental letter dated March 20, 2006, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 9, 2007.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC

    Docket No. 50-220, Nine Mile Point Nuclear Station, Unit No. 1, 
Oswego County, New York.
    Date of application for amendment: October 19, 2006, as 
supplemented by letter dated January 5, 2007.
    Brief description of amendment: The amendment revises the 
Surveillance Requirement (SR) in Technical Specification (TS) 4.1.1.c, 
``Scram Insertion Times,'' to modify the conditions under which scram 
time testing (STT) of control rods is required, and to add a 
requirement to perform STT on a defined portion of control rods, at a 
specified frequency, during the operating cycle. The amendment also 
revises the SR in TS 4.1.7.c, ``Minimum Critical Power Ratio (MCPR),'' 
to add a requirement to determine the MCPR operating limits following 
completion of control rod STT per TS 4.1.1.c.
    Date of issuance: March 15, 2007.
    Effective date: March 15, 2007.
    Amendment No.: 193.
    Facility Operating License No. DPR-63: Amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: December 5, 2006 (71 FR 
70562) The supplemental letter dated January 5, 2007, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 15, 2007.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC

    Docket No. 50-311, Salem Nuclear Generating Station, Unit No. 2, 
Salem County, New Jersey.
    Date of application for amendment: April 6, 2006.
    Brief description of amendment: The amendment changed the Technical 
Specifications (TSs) to reduce the maximum allowable reactor power 
level when two main steam safety valves are inoperable.
    Date of issuance: March 7, 2007.
    Effective date: As of the date of issuance and shall be implemented 
prior to restart from the steam generator replacement outage.
    Amendment No.: 259.
    Facility Operating License No. DPR-75: The amendment revised the 
TSs and the License.
    Date of initial notice in Federal Register: November 7, 2006 (71 FR 
65144).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 7, 2007.
    No significant hazards consideration comments received: No.

R.E. Ginna Nuclear Power Plant, LLC

    Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County, 
New York.
    Date of application for amendment: May 1, 2006, as supplemented by 
letter dated November 3, 2006.
    Brief description of amendment: The amendment re